05000416/LER-2018-007, Potential Loss of Safety Function (Residual Heat Removal) and System Actuation Caused by Inadvertent Valve Opening
| ML18177A329 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/26/2018 |
| From: | Emily Larson Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GNRO-2018/00032 LER 2018-007-00 | |
| Download: ML18177A329 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4162018007R00 - NRC Website | |
text
c~Entergy GNR0-2018/00032 Jun 26, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Operations, Inc.
P. 0. Box756 Port Gibson, MS 39150 Eric A. Larson Site Vice President Grand Gulf Nuclear Station Tel. (601) 437-7500 10CFR50.73
SUBJECT:
Licensee Event Report 2018-007-00, Potential Loss of Safety Function (Residual Heat Removal) and System Actuation Caused by Inadvertent Valve Opening
Dear Sir or Madam:
Grand Gulf Nuclear Station, Unit 1
- Docket No. 50-416 License No. NPF-29 Attached is Licensee Event Report 2018-007-00. This report is being submitted in accordance with 10CFR50.73(a)(2)(v)(B) as an event or condition that could have prevented fulfillment of a safety function, and under 10CFR50.73(a)(2)(iv)(A) for the Emergency Core Cooling System (ECCS) and associated Diesel Generator System Actuations.
This letter contains no new commitments. If you have any questions or require additional information, please contact Douglas Neve at 601-437-2103.
Sincerely, Eric A. Larson Site Vice President Grand Gulf Nuclear Station EAL/jw
Attachment:
Licensee Event Report 2018-007-00 cc: see next page
GNR0-2018/00032 Page 2 of 2 cc:
NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S. Nuclear Regulatory Commission ATTN: Ms. Lisa M. Regner Mail Stop OWFN 8 81 Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Mr. Kriss Kennedy, NRR/DORL (w2)
Regional Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511
GNR0~2018/00032 Attachment Licensee Event Report 2018-007 -00
- - U;S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017)
Estimated burden. per response to comply with this mandatory collection request: 80 1*
hours. Reported lessons learned are incorporated into the licensing process and fed
~BRE"G(lt LICENSEE EVENT REPORT (LER) back to industry. Send comments regarding burden estimate to the Information Services
~r:;
.,,)!_
Branch (T-2 F43), U.S. Nuclear Regulatory.Commission, Washington, DC 20555-0001, 0~
o~
(See Page 2 for required number of digits/characters for each block) or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of
/!!'
()
Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management 0
and Budget, Washington, DC 20503: If a means used to impose an information collection
~
.~:
(See NUREG-1022, R.3 for instruction a_nd guidance for completing this does not display a currently valid OMB control number, the NRC may not conduct or
~
sponsor, and a person is not required to respond to, the information collection.
~~
o" form
_Q****"'~ -
http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 1. FACILITY NAME*
- 2. DOCKET NUMBER
- 13. PAGE Grand Gulf Nuclear Station, Unit 1 05000 416 1 OF 3
- 4. TITLE Potential Loss of Safety Function (Residual Heat Removal) and System Actuation Caused by Inadvertent Valve Opening
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I SEQUENTIAL J REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER
_ NO. _
MONTH DAY YEAR NIA D-5000.N/A FACILITY NAME DOCKET NUMBER 05 01 2018 2018-007-00 06 26 2018 NIA 05000 N/A
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMEN'IS OF 10 CFR §: (Check all that apply)
- 9. OPERATING MODE D 20.2201(b)
D 20.2203(a)(3)(i+*)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 5 D 20.2201(d)
D 20.2203(a)(3)(ii)
D 50. 73(a)(2)(ii)(B)
D 50. 73(a)(2)(viii)(B)
D 20.2203(a)(1 >
D 20.2203(a)(4)
D 50. 73(a)(2)(iii)
D 50. 73(a)(2)(ix)(A)
- - D-20.2203(a)(2)(i) 50.36(c)(-1)(i)(A)
.cgj 50.73(a)(2)(iv)(A)
- - D 50.73(a)(2)(x)
D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
- 10. POWER LEVEL D 20.2203(a)(2)(iii)
D 50.36(c)(2) cgJ 50.73(a)(2)(v)(B)
D 73.71(a)(5)
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1) 0 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50:73(a)(2)(i)(C) 00THER Specify in Abstract below or in NRC Form 366A t2. *LICENSEE CONTACT FOR THIS-LER LICENSEE CONTACT I TELEPHONE NUMBER (Include Area Code)
Douglas Neve I Manager, Regulatory Assurance (601) 437-2103 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX A
NIA NIA NIA NIA NIA NIA NIA NIA NIA I
- 14. SUPPLEMENTAL REPORT E)_CPECTED
- 15. EXPECTED MONTH DAY YEAR O YES (If yes, complete 15. EXPECTED SUBMISSION DA TE) [g} NO SUBMISSION DATE NIA NIA NIA ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On May 1, 2018, Grand Gulf Nuclear Station Instrument & Control technicians started a Reactor Vessel Water Level Transmitter Calibration Surveillance. At 1551, the technicians inadvertently opened the low pressure isolation valve instead of the equalization valve. This resulted in a decrease in sensing line pressure, which appeared as a low water level signal to the transmitters. As a result, Division 1 Emergency Core Cooling System initiated, and Shutdown Cooling
- *was-isolated. All systems-responded as expected. After a 5 degree Frise *in *local Reactor*Coolant*System *(RCS) temperature, Operations restored Shutdown Cooling.
The technicians inadvertently opened the wrong valve because of a failure to use human performance tools. In addition, Operations did not implement an adequate risk mitigation strategy for the surveillance. Both of these causes will be corrected by revising surveillances which involve replacement, calibration, or maintenance on consequential transmitters, to include hardened barriers.
T-he-safety--consequences-ofthis event-were low, as -Shutdown-Cooling-was-restored after only-a *slight-rise-in local-RC-S temperature.
Page 1 of 3 A.
PLANT CONDITIONS PRIOR TO THE EVENT APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
YEAR I
- 3. LER NUMBER SEQUENTIAL NUMBER 2018-007-00 I REV.
NO.
Mode 5, Reactor Coolant System (RCS) Temperature 87 F, Residual Heat Removal (RHR) [BO] Train "A" aligned for Shutdown Cooling No inoperable structures, components, or safety systems contributed to this event.
B.
DESCRIPTION
on-May 1, 2018, Grand GulfNuclearStation *instrument & Control (1-&C) technicians started a Reactor Vessel Water Level Transmitter Calibration Surveillance. At 1551, the technicians inadvertently opened the
'low pressure isolation valve instead of the equalization valve. This resulted in a decrease in sensing line pressure, which appeared as a low water level signal to the transmitters. As a result, Division 1 Emergency Core Cooling System (ECCS) initiated, and Shutdown Cooling was isolated. All systems responded as expected, including RHR "A", which automatically aligned to inject into the Reactor. The RHR "A" suction source remained from the spent fuel pool, and thus there was not a net change in RCS inventory. The ECCS actuation resulted in minimal flow from the RHR "A" pump through the "RHR "A" Heat Exchangers.
After a -5 degree "F rise in *1ocal Reactor *coolant-System (RCS) temperature, -Operations realigned RHR"A" from Injection to Shutdown Cooling mode.
C. REPORTABILITY
This event is being reported under 1 OCFR50. 73(a)(2)(v)(B), as an event or condition that could have prevented fulfillment of a safety function (Residual Heat Removal), and under 10CFR50.73(a)(2)(iv)(A) for) the Emergency Core Cooling System (ECCS) and associated_ Diesel Generator [EK] System Actuations.
The event was initially reported under 1 OCFR50.72(b)(2)(iv)(A) for the ECCS and associated Diesel Generator System Actuations and 50.72(b)(3)(v)(B) for the potential loss of safety function (Residual Heat Removal) on May 1, 2018, via Event Report 53374.
D. CAUSE
The technicians inadvertently opened the wrong valve because of a failure to use human performance
.tools. J&C technicians.failed to.use various.human.performance tools,.including.procedure adherence, operating experience, questioning attitude, verification/validation, peer check, and self-check. Furthermore, flagging or robust barriers were not used in valve manipulations. In addition, Operations did not implement an adequate risk mitigation strategy for the surveillance. This task had been previously categorized as high impact; however, it was screened as medium during this event.
E. CORRECTIVE ACTIONS
The following actions are completed or planned.
Page 2 of 3 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020 (4-2017) v~Rro11<.q LICENSEE EVENT REPORT (LER)
~-:;
"'o G
~
CONTINUATION SHEET 0
~.
i
~1,~
- o~
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/readinq-rm/doc-collections/nuregs/staff/sr1022/1'3!)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME
- 2. DOCKET YEAR Grand Gulf Nuclear Station, Unit 1 05000 416 Completed:
I
- 3. LER NUMBER SEQUENTIAL NUMBER 2018-007-00 I The Control Room was notified. Operations restored Shutdown Cooling by realigning "A Residual He.at Removal, after only _a.slight.ris.e.in.local RCS temperature.
The qualifications were removed from the involved individuals.
Planned actions included in the corrective action program which may be changed in accordance with the program:
Revise surveillances, which involve replacement, calibration, or maintenance on transmitters that *have the ability to trip the plant, trip the turbine, or result *in a sa*fety system actuation, to include hardened barriers.
F.
SAFETY SIGNIFICANCE
REV.
NO.
The actual consequences of this event were an unplanned initiation of Division 1 ECCS and a temporary loss of Shutdown Cooling. There were no other actual consequences to the general safety of the _public, nuclear safety, industrial safety, and radiological safety for this event.
The potential consequence to the general safety of the public, nuclear safety, industrial safety, and radiological safety for this event is a continued rise in the RCS temperature until it reached 212 degrees F, and began to boil. RHR "A" was realigned, in accordance with plant procedures, to provide Shutdown Cooling after only a 5 degree F rise in RCS temperature. In addition, RHR "B" was also available.
Therefore, the residual heat removal safety function was never lost. Since the time to boil at the start of the
.event was.greater than 7.hours,.and there were multiple systems.available to.remove.residual.heat, the.risk.
associated with this event is low.
Based on the above, Entergy has determined that it did not result in an actual loss of safety function.
Therefore, in accordance with the guidance provided in NE1 99-02, Revision 7, Regulatory Assessment Performance Indicator Guideline, Section 2.2, Mitigating Systems Cornerstone, Sub-Section, Safety System Functional Failures, Page 30, Lines 27 through 30; this condition will not be counted as a safety system functional failure against Performance Indicator MS05, Safety System Functional Failures.
G. PREVIOUS SIMILAR OCCURRENCES An internal Operating Experience search of the Corrective Action system was performed for the previous fifteen yearsr and no similar events could be found.